ML15110A129

Correction to the Emergency Planning Exemptions Safety Evaluation

ML15110A129
Person / Time
Site:	Crystal River
Issue date:	04/22/2015
From:	Michael Orenak Plant Licensing Branch IV
To:	Hobbs T Duke Energy Florida
Wengert T
References
TAC MF2981
Download: ML15110A129 (5)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 22, 2015 Mr. Terry D. Hobbs Decommissioning General Manager Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, FL 34428-6708

SUBJECT:

CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT - CORRECTION TO THE EMERGENCY PLANNING EXEMPTIONS SAFETY EVALUATION (TAC NO. MF2981)

Dear Mr. Hobbs:

By letter dated March 30, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15058A906), the U.S. Nuclear Regulatory Commission (NRC) approved exemptions from specific requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.47, "Emergency plans," and Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," to 10 CFR Part 50 for Crystal River Unit 3 Nuclear Generating Plant (CR-3). The approval was in response to the Duke Energy Florida, Inc. (DEF), application for exemptions dated September 26, 2013, "Crystal River Unit 3 -

License Amendment Request #315, Revision 0, Permanently Defueled Emergency Plan and Emergency Action Level Scheme and Request for Exemption to Certain Radiological Emergency Response Plan Requirements Defined by 10 CFR 50," supplemented by letters dated March 28, May 7, May 23, and August 28, 2014 (ADAMS Accession Nos. ML13274A584, ML14098A072, ML14139A006, not publically available, and ML14251A237, respectively).

The exemptions eliminated the requirements to maintain an offsite radiological emergency plan and reduce the scope of onsite emergency planning activities at CR-3 based on the reduced risks of accidents that could result in an offsite radiological release at a decommissioning nuclear power reactor.

Subsequent to issuing the exemptions, an error was identified in the supporting safety evaluation (SE). Your staff pointed out that an identified procedure, EM-202, "Duties of the Emergency Coordinator," is incorrect on page 15 of the SE. The properly identified procedure should be EM-502, "Conduct of the Emergency Coordinator," which is the updated version of EM-202. DEF referenced EM-202 in the original September 26, 2013, letter, but in the May 23, 2015, supplement, DEF described the update to EM-502. The incorrectly identified procedure has no impact on the SE or the SE conclusion that the emergency planning requirements for CR-3, as modified by the exemptions described in the SE, would provide: (1) an adequate basis for an acceptable state of emergency preparedness; and (2) in conjunction with arrangements made with offsite response agencies, reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency at CR-3.

T. Hobbs If you have any questions, please contact me at 301-415-3229 or via e-mail at Michael. Orenak@nrc.gov.

Sincerely,

~ ~ (}wl Michael D. Orenak, Project Manager Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302

Enclosure:

Corrected page 15 of the safety evaluation cc w/enclosure: Distribution via Listserv

CORRECTED SAFETY EVALUATION PAGE 15 DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302

The licensee stated that, in accordance with License Amendment No. 239, issued December 27, 2011 (Reference 30), and License Amendment No. 241, issued June 26, 2012 (Reference 31 ), CR-3 will complete replacement of the auxiliary building cask handling crane with a single failure proof crane prior to moving a spent fuel shipping cask. The NRC staff finds that the qualification and operation of the CR-3 cask handing crane as single-failure-proof handling system satisfies the conditions assumed in the staff's analysis presented in NUREG-1738 with respect to protection from potential cask drop events.

IDC #2 states: Procedures and training of personnel will be in place to ensure that onsite and offsite resources can be brought to bear during an event.

IDC #3 states: Procedures will be in place to establish communication between onsite and offsite organizations during severe weather and seismic events.

IDC #4 states: An offsite resource plan will be developed which will include access to portable pumps and emergency power to supplement onsite resources. The plan would principally identify organizations or suppliers where offsite resources could be obtained in a timely manner.

The licensee described how the proposed CR-3 PDEP; the associated implementing procedures EM-502, "Conduct of the Emergency Coordinator," and EM-503, "Conduct of the Emergency Mitigation Coordinator"; and the Off-site Support Directory would provide for access to offsite resources, including provisions for training, communications, and coordination to obtain offsite resources. The NRC staff concludes that the licensee has adequate procedures to satisfy the conditions assumed in the NUREG-1738 analysis regarding effective use of onsite and offsite resources to respond to events affecting the SFP.

IDC #5 states: Spent fuel pool instrumentation will include readouts and alarms in the control room (or where personnel are stationed) for spent fuel pool temperature, water level, and area radiation levels.

The licensee stated that independent narrow range SFP level indication is available in the main control room for both the "A" SFP and the "B" SFP, which are normally connected through an open gate. In addition, the facility is equipped with control room alarms for high and low SFP level, SFP high temperature via the plant computer, and high SFP area radiation levels.

Therefore, the NRC staff finds that the licensee will maintain adequate SFP monitoring instrumentation to satisfy the conditions assumed in the NUREG-1738 analysis regarding monitoring events affecting the SFP.

IDC #6 states: Spent fuel pool seals that could cause leakage leading to fuel uncovery in the event of seal failure shall be self-limiting to leakage or otherwise engineered so that drainage could not occur.

The spent fuel storage area contains two gates. One gate isolates the "A" SFP from the "B" SFP and one gate isolates the "B" SFP from the Cask Area. The licensee stated that the gates are not normally installed and the bottom of each SFP gate opening is located about 1 foot above fuel stored in rack modules. Therefore, the configuration of the gate openings limits the leakage from the storage pools. The NRC staff finds that the described design

..* ML 1511OA129 OFFICE NRR/DORULPL4-2/PM NRR/DORULPL4-2/LA NRR/DORULPL4-2/BC NRR/DORULPL4-2/PM NAME MOrenak PBlechman MKhanna MOrenak DATE 4/21/15 4/21/15 4/22/15 4/22/15